Tritium, T or 3H, is a radioactive isotope of hydrogen. Tritium is a crucial component of thermonuclear weapons and tritium gas is used in U.S. nuclear warheads to enhance explosive yield. The radioactive decay rate of tritium is 5.5% per year, resulting in a half-life of a little over 12 years. Consequently, it is desirable to provide a stable, safe and efficient supply of tritium for defense and research purposes.
Tritium is separable from lighter isotopes, but only by very tedious, expensive methods. An alternative to tritium isolation is tritium production in which other elements are transmutated to tritium through neutron capture. For example, tritium may be produced by thermal neutron capture by 6Li which decays to tritium and helium. Commercial power and research nuclear reactors produce an abundance of thermal neutrons which might potentially be used in producing tritium through neutron capture transmutation reactions. Accordingly, a lithium-containing compound may be disposed in the core of a nuclear reactor to produce tritium.
Particularly suitable lithium-containing compounds for tritium production are the lithium aluminum oxides, LiAlO2 and LiAl5O8, also referred to as “lithium aluminates,” which have high atom percents of lithium and have high melting points (respectively about 1610° C. and 1900° C.). Lithium aluminum oxide may be provided in the form of minute spherical particles as is taught in U.S. Pat. No. 4,405,595 to Yang, assigned to the assignee of the present invention.
One drawback of tritium production using commercial power or research reactors is that tritium is both radioactive and it presents substantial handling difficulties. Like other hydrogen isotopes, tritium diffuses easily through many materials, including the metal cladding used in convention tritium production elements. Early efforts to address tritium permeation focused on encasing the production material in tritium-impermeable container. As such method, described in the above-mentioned patent to Yang, involves coating a particulate material, such as lithium aluminum oxide, with a tritium-impermeable shell. In particular, lithium aluminum oxide particles are coated with a tristructural isotopic (TRISO) coating similar to that used to coating nuclear fuel particles in research-style reactors. Such coatings consist of a layer of porous carbon, a layer of an isotropic dense carbon, a layer of silicon carbide, and a layer of an isotropic dense carbon. However, such previously-known arrangements require substantial post-irradiation processing.
Likewise, U.S. Pat. No. 4,597,936 to Kaae, assigned to the assignee of the present invention, discloses a lithium-containing neutron target particle for breeding tritium within the core of a nuclear reactor, including a central core formed of a stable lithium-containing compound, a surrounding buffer layer, and an outer tritium-impermeable silicon carbide coating. The core is initially sealed with an inner sealing layer of pyrolytic carbon and an outer sealing layer of stoichiometric zirconium carbide.
Tritium production elements, such as described in the above patents, generally are not used in commercial power reactors due to the low relative loading of lithium. For this reason, until 1993, most tritium production in the United States was undertaken at the Savannah River production reactor, under the auspices of the U.S Department of Energy.
Since about 2005, efforts have been underway to develop tritium production rods for use in commercial power reactors, such as those belonging to the Tennessee Valley Authority. The tritium production rods employed in those efforts are referred to as “TPBARs” or “tritium producing burnable absorber rods” and comprise stacks of lithium aluminate disposed within a stainless steel cladding. However, due to the high permeability of tritium in stainless steel, such designs have required increased emphasis on tritium extraction from the reactor coolant, and also pose a risk to conventional zirconium alloy clad fuel assemblies in the power reactor, as described below.
Efforts to employ barriers to reduce hydrogen and tritium permeation in conventional nuclear fuel cladding also are known in the art. For example, it has long been known that zirconium alloy cladding, used in most commercial power reactors, is susceptible to hydrogen embrittlement, wherein high levels of hydrogen permeation into the cladding thickness can reduce the structural integrity of the cladding. Accordingly, prior art patents, such as U.S. Pat. No. 5,026,516 to Taylor and U.S. Pat. No. 5,341,407 to Rosenbaum describe the use of a pure zirconium barrier within a zirconium alloy tube to reduce hydrogen uptake and embrittlement from within the fuel rod. Likewise, U.S. Patent Application Pub. No. 2009/0238322 to Liu discloses a fuel assembly for a nuclear reactor wherein the fuel rods include hollow gas absorber structures, referred to as “getters”, which are disposed within the fuel rods to absorb and retain hydrogen and tritium. The presence of high tritium levels outside the cladding, e.g., released by TPBARs located within the reactor, may lead to cladding embrittlement from the outside in.
U.S. Patent Application Pub. No. 2009/0032178 to Feinroth discloses a nuclear fuel cladding intended to overcome the disadvantages of zirconium-based fuel claddings, and in particular, the exothermic corrosion of zirconium alloys that occurs when a heated fuel rod is exposed to air or steam. Feinroth describes a nuclear fuel cladding comprising a multi-layered ceramic tube having an inner layer of high purity beta phase stoichiometric silicon carbide, a central composite layer of continuous beta phase stoichiometric silicon carbide fibers, and an outer layer of fine-grained silicon carbide.
Tritium permeation also has been hypothesized to pose problems for advanced reactor designs, such as fusion reactors, in which deuterium and tritium are combined in a fusion reaction to generate helium. For example, Sandia National Laboratories (Albuquerque, N. Mex.) describes, in a publication entitled “Silicon Carbide Permeation Barrier for Steel Structural Components”, a tritium permeation barrier for steel structural components of a fusion reactor, in which a silicon carbide coating and a compliant foam interlayer are disposed within a ferritic steel tube. That publication is directed to addressing tritium permeation barriers in the context of nuclear fusion reactor components, and is not concerned with tritium production in conventional power reactors.
In view of the above-noted drawbacks of previously-known systems, it would be desirable to provide a tritium production element that experiences reduced permeation, thereby permitting safe and economical production of tritium in conventional nuclear power reactors.
Due to the long lead times needed to obtain regulatory approval for new reactor designs and due to the relatively short half-life of tritium, it would be particularly desirable to provide methods and apparatus for producing tritium suitable for use with existing reactor facilities and modes of operation.
Further, it would be desirable to provide methods and apparatus for producing tritium that is compatible with zirconium alloy based nuclear cladding used in current commercial power reactors.